ML20206A615

Summary of Facility Changes,1986

ML20206A615
Person / Time
Site:	North Anna
Issue date:	04/01/1987
From:	Stewart W VIRGINIA POWER (VIRGINIA ELECTRIC & POWER CO.)
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
87-118, NUDOCS 8704080104
Download: ML20206A615 (37)

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VIRGINIA ELECTRIC AND POWEH COMPANY I

RIciswoNn,VINGINIA 20261

w. L. stewaar Vaca Passinawr

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April 1, 1987 U. S. Nuclear Regulatory Commission Serial No.87-118 Attention: Document Control Desk NAPS /JHL Washington, D. C. 20555 Docket Nos. 50-338 50-339 License Nos. NPF-4 i NPF-7 )

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Gentlemen:

VIRGINIA ELECTRIC AND POWER COMPANY NORTH ANNA POWER STATION UNITS 1 AND 2

SUMMARY

OF FACILITY CHANGES Pursuant to 10 CFR 50.59(b), attached is the Summary of Facility Changes for North Anna Units 1 and 2. A summary of the safety analysis is provided for each facility change.

Very truly yours, m

C a W. L. Stewart Attachment cc: U. S. Nuclear Regulatory Commission 101 Marietta Street, N. W.

Suite 2900 Atlanta, Georgia 30323 Mr. J. L. Caldwell NRC Senior Resident Inspector North Anna Power Station 0704080104 870401 PDR R

ADOCK 05000338 PDR [).

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I DC 81-20B UNIT 2 NUREG 0696 SHORT TERM I&C PROJECT REMOTE MULTIPLEXER INSTALLATION Description NUREG-0696 " Functional Criteria for Emergency Response Facilities" establishes the requirement that all nuclear power plant licensees install facilities which include a Technical Support Center (TSC),

Onsite Operational Support Center (OSC), Emergency Operation Facility (EOF), Safety Parameter Display System (SPDS) and a Nuclear Data Link (NDL).

These facilities provide the following:

a) Plant information during normal operation and informatica necessary to assess plant conditions following an accident.

b) Plant safety status c) Improved emergency response and more effective emergency management d) Transmission of more accurate information to Federal, State and Jocal authorities In order to satisfy the criteria established in NUREG-0696, a complete and independent Data Acquisition System manufactured by Validyne Inc.

(model HD310) was installed. This package provides instruction for the mounting, installation and wiring of the front-end remote multiplexer units for the DAS. The system is capable of collecting plant information and transmitting it to the ERF and display systems.

Summary of Safety Analysis The NUREG-0696 Short Term I&C Project-Remote Multiplexer Installation does not create an "unreviewed safety question" as defined in 10CFR50.59. This modification installs the front-end multiplexer units  ;

of the Data Acquisition System which collects and transmits reliable i plant data and information to the Emergency Response Facilities. The front-end multiplexer units, mounting arrangements, cabling and wiring to the units will be Q.A. Category I Class 1E. The multiplexer units have been qualified as a Class 1E isolation device, therefore connections to existing Class 1E equipment and circuits do not degrade the operation of safety-related components or protection circuits and will not interfere with the operation of control room monitoring devices and plant system functions.

DC 81-20K UNIT 2 SUPPLEMENTAL S.P.D.S. INPUTS INSTALLATION Description Supplement I to NUREG-0737 " Requirements for Emergency Response Facilities" requires that nuclear power stations upgrade their facilities to include a Safety Parameter Display System (SPDS) in the control room for each unit to provide a concise display of critical plant variables to the control room operators to aid them in rapidly and reliably determining the safety status of the plant.

In order to satisfy the requirements of Supplement I to NUREG-0737 and guidelines established in NUREG-0696, additional parameters which were not covered under DC-81-20B were input into the SPDS. A multiplexing system manufactured by Validyne Inc. (Model HD310), and installed by DC-81-20B, D, and DC-81-20I collects the pertinent plant data.

Data displayed continuously by the dedicated primary display of the SPDS includes but is not limited to:

1. Reactivity control

2. Reactor core cooling and heat removal from primary system

3. Reactor coolant system integrity

4. Radioactivity control

5. Containment integrity This design change package provides instructions for wiring additional inputs into the Data Acquisition System (DAS) needed to support the SPDS displays. Additional multiplexers, related multiplexer hardware and wiring (fiber optic cables and power cables) were required to implement the additional plant variable inputs.

Summary of Safety Analysis This equipment is used to provide additional critical plant variable inputs to the SPDS and to the control room personnel to quickly assess the status of safety systems during abnormal and emergency conditions.

The additional inputs to the SPDS are an improvement in station design and emergency response capability, providing the control room personnel with critical plant information necessary to assess the status of the safety systems.

This modification is an improvement in station response capability.

Additional critical plant variables are displayed on the SPDS and the displays assist control room personnel in the detailed analysis of abnormal plant conditions and in the implementation of emergency response plans.

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DC 82-14B UNIT 2 CLASS 1E TRANSMITTER REPLACEMENT Description Regulatory Guide 1.97 (RG-1.97) " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" requires that sensing instrumentation which provides key plant variables to operations personnel in the control room and designated Category I shall be seismically and environmentally qualified.

A review of the instrumentation required to satisfy RG-1.97 identified instruments which required replacement with environmentally qualified equipment.

Summary of Safety Analysis The replacement transmitters are seismically and environmentally qualified, provide Category I indication in the main control room and their installation does not affect safety equipment or the operation of safety systems previously evaluated.

The replacement transmitters are environmentally qualified, capable of surviving the accident environment and therefore, will provide more reliable Category I indication of various plant variables and system l functions during and following a design basis event.

The replacement transmitters provide the operator with Category I indication to assess the operation of safety systems and do not change safety limits or setpoints defined in the Technical Specifications.

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DC 83-33 UNIT 2 CLASS lE SOV REPLACEMENT l

Description The Code of Federal Regulations, Section 10CFR50.49 " Environmental Qualification of Electric Equipment Important to Safety" requires that Category I, Class IE electrical equipment shall be environmentally qualified and must be capable of functioning under postulated accident conditions. A review conducted in accordance with guidelines established in IE-Bulletin 79-OlB and NUREG-0588 identified several solenoid valves (SOVs) (used for containment isolation functions) that did not have sufficient documentation available to verify their l qualification.

4 The SOVs which were identified were removed and replaced with SOVs which l'

are environmentally qualified in accordance with IEEE-323-1974 and seismically qualified in accordance with IEEE 344-1975. The replacement components are similar in design to those being replaced and were installed in the same location, mounted in accordance with j manufacturer's instruction, and reterminated with qualified cable seal j connectors.

Summary of Safety Analysis The replacement SOVs which are seismically and environmentally qualified, replaced existing S0Vs on containment isolation valves and j their installation does not change the operation of the valves or affect 1 safety equipment or the operation of safety systems.

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The replacement SOVs are environmentally qualified, capable of surviving l the accident environment, and are a direct replacement of existing components.

This modification involved a one for one repincement of existing SOVs for various containment isolation valves with components which are environmentally qualified, and does not change safety limits or limiting safety system settings discussed in the technical specifications.

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DC 84-18 UNIT 2 VALVE MONITORING SYSTEMS: PORV AND PRESSURIZER SAFETY VALVES Description The Valve Monitoring System (VMS), presently installed for the Pressurizer Power Operated Relief Valves (PORVs) and the Safety Valves SVs), was not seismically and environmentally qualified as required by USNRC Regulatory Guide 1.97 Rev. 3.

Each PORV, (PCV-2455C and PCV-2456) had two non-qualified limit switches to monitor open/ closed position. These limit switches were removed and in their place (4) four new qualified limit switches for redundant open/ closed position monitoring were installed.

Non-qualified acoustic monitors were located on the discharge of SVs SV-2551A, SV-2551B and SV-2551C. These non-qualified monitors and the preamplifiers associated with these monitors were removed and, in their place, new qualified acoustic monitors and preamplifiera to monitor SV

, position were installed as required by R. G. 1.97.

Summary of Safety Analysis The replacement limit switches and acoustic monitors are environmentally and seismically qualified. They provide indication to the operator to monitor station operation, and their installation does not affect safety equipment or the operation of safety systems. The replacement limit switches and acoustic monitors are environmentally qualified, capable of surviving the accident environment and, therefore, will provide more reliable indication of station operation during a design basis event.

The margin of safety as defined in the basis for any Technical Specification is not reduced. The replacement limit switches and acoustic monitors are environmentally and seismically qualified instruments which provide the operator with information to assess the i operation of safet/ systems.

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DC 84-42 UNIT 2 APPENDIX R - SPURIOUS OPERATIONS OF HIGH/ LOW PRESSURE BOUNDARY VALVES i

l Description j 10CFR50, Appendix R, Section III, G.2, requires that certain means be l

provided for ensuring safe operation of redundant trains of systems necessary to achieve and maintain safe shutdown conditions. The Reactor

]j Coolant System High/ Low Pressure valve pairs must be prevented from t spuriously operating as a result of a fire affecting the cables of these 3 valves.

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Spurious operations may occur when a voltage or signal from one circuit imparts a voltage on an adjacent circuit of a similar voltage class.

This is only possible when a fire burns away the insulating material on

! circuits in proximity. This type of operation includes the case of one de-energized circuit becoming energized by shorting to an external

source of electrical power through conductor to conductor shorts.

l Since the valves are three phase, their feeder cables are not j susceptible to hot shorts (by definition). However, the fire that may

! cause the valves to spuriously opera.e also has been postulated to open the 3 phase cable to the valve. This open circuit causes the valve to l " fail as is" which could be open. The valve thus car.not be closed, so protection is provided for the 125 V de solenoid operated valve.

! Protection from hot shorts for 125V de activated valves has been i provided by routing the associated cables entirely in dedicated conduits I and providing two switches in series of each circuit (one each in the j Control Room Benchboard 2-1 and Emergency Switchgear Room Appendix R i Isolation Panel). The steel conduit provides protection from external hot shorts and either of the two switches provides protection from internal shorts.

Summary of Safety Analysis This modification prevents the occurrence of spurious malfunction during

fire conditions, therefore it reduces the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety. In addition, termination of the cables to the valves was performed during unit shutdown or refueling; therefore, no additional increase in the probability of an accident or malfunction of equipment important to safety was created. This design change enhances the ability to maintain reactor coolant inventory by eliminating breaches in the system as a result of a fire.

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f DC 84-42 UNIT 2 APPENDIX R - SPURIOUS OPERATIONS OF HIGH/ LOW PRESSURE BOUNDARY VALVES Summary of Safety Analysis (cont)

The materials used in this modification are totally compatible (relative to design, quality, and functional requirements) with original

' equipment. This modification provides a method of assuring the High/ Low i Pressure Boundary can be maintained during fire. Therefore, this modification reduces the possibility for an accident or malfunction of a different type than previously evaluated.

The design change enhances the ability to maintain reactor coolant inventory by preventing spurious operation from occurring in the event of a fire. This design change does not affect the basis of any Technical Specification requirements.

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! DC 84-48 UNIT 2 REG. GUIDE 1.97 TRANSMITTER MODIFICATION

~i Description j The Refueling Water Storage Tank Level transmitters (LT-QS 200 A, B C &

D) were not seismically and environmentally qualified as required by IEEE 323-1974, IEEE 344-1975, 10CFR50.49 and NUREG 0588, and R.G. 1.97, <

i Rev. 3 " Instrumentation for Light-Water-Cooled Nuclear Power Plant to l

) Assess Plant and Environs Conditions During and Following an Accident". l They must be qualified in order to meet the category I requirements of R.G. 1.97, Rev. 3 (Item A-10). The present Refueling Water Storage Tank 1

level transmitters were replaced with environmentally and seismically qualified level transmitters which meet the qualification requirements for electrical class IE installations (IEEE 323-1974, IEEE 344-1975, 10CFR50.49 and NUREG 0588, Category 1 per R.G. 1.97 Rev. 3). The

transmitters were installed at the same location in new seismically supported enclosures.

i Summary of Safety Analysis ,

J The replacement transmitters are environmentally and seismically j qualified to provide indication to the operator to monitor station 1 operation, and their installation does not affect safety equipment or

,! the operation of safety systems.

I The replacement transmitters are environmentally qualified, capable of surviving the accident environment and therefore, will provide more j reliable indication of station operation during a design baais event.

1 i The replacement transmitters are environmentally and seismically i qualified instruments which provide the operator with information to j assess the operation of safety systems.

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4 DC 81-15 UNIT 1 and 2

! TECHNICAL SUPPORT CENTER FACILITIES REMOVAL  !

Description f

i i i In accordance with the requirements of NUREC-0696 and 10CFR50, Appendix >

l E Article IV.E.8 a Technical Support Center (TSC) has been established.

! In order to provide space for the TSC, existing electrical, building i

service and structural facilities in the old Machine Shop area of the

] Service Building have been removed in accordance with the specific j details and instructions of this DCP.

) The TSC facilities installation was performed by DCP's 81-15, 81-17, j 81-18 and 81-19.

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Summary of Safety Analysis

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There is no design basis accident that could be initiated by the Machine Shop facilities removal. This DCP modified some of the existing fire protection piping to hose racks while always maintaining provisions for fire protection. The consequences of an accident previously evaluated

in the UFSAR has not been increased. The Control and Relay Rooms, to which the Machine Shop is structurally interfaced, have not been l degraded due to the Machine Shop demolition. The Machine Shop facilities removal does not involve modifications to safety related i equipment previously evaluated in the UFSAR or systems that may create l accidents or malfunctions of a different type than previously evaluated j in the UFSAR. The margin of safety as defined in the basis for any

Technical Specification is not reduced. ,

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i DC 81-16 UNIT 1 and 2 i i TECIINICAL SUpp0RT CENTER RADIATION MONITORING SYSTEM -

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! Dencription i

I l The design basis for the TSC radiation monitors is provided in j NUREG-0696 " Functional Criteria for Emergency Response Facilities,"

Section 2.6, liabitability which states that these systems shall t

continuously indicate radiation dose rates and airborne radioactivity concentrations inside the TSC while it is in use during an emergency.  !

! The structural design of the TSC provides radiation shielding for personnel working in the TSC from radiation outside the TSC. Radiation monitoring equipment was installed to provide exposure information to i plant pernonnel.

A radiation monitoring system meeting the requiren.ents of NUREC-0696,

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i Section 2.6 has been installed in the Technical Support Center.

monitoring system includes an Eberline PING-3B particulate, iodine, and The .,

I j noble gas monitor, two Eberline EC4-X area monitora, an Eberline RIE-S l l remote alarm panel, interconnecting cable, and stainless nteel sampic l j lines, i i  !

j Summary of Safety Analysis l

]' The TSC radiation monitors are not safety-related nor do they affect any l safety-related equipment. This design change does not create an l "unreviewed staf ety question" as defined in 10CFR50. Specifically j i

n. This change does not increase the probability of occurrence or the i consequences of an accident or malfunction of equipment important  :

to nafety and previously evaluated in the UI"iAR. The new equipment I is not safety-related. i

b. The possibility of an accident or malfunctiot, of a different type l than any previounty evaluated in the UFSAR is not created. The j

! monitorn are not part of any safety-reinted nyitens, structuren, or  ;

j component require (! to mitigate an accident.

i i l c. The margin of anfety as defined in the basis for any Technical  !

Specification in not reduced. The radiation monitors are for TSC

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j personnel radiation protection in a post-accident environment.

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DC 81-17 UNIT 1 and 2 l TECilNICAL SUPPORT CENTER FACILITIES INSTALLATION l  !

Description The requirements of NUREG 0696, 10CFR50, Appendix E. Article IV.E.8 and Appendix A GDC 19 nddress the establishment of a habitable Technical Support Center (TSC) facility. In order to provide a radiologically  :

habitable facility, additional facilitica have been provided. .i The TSC has been installed in the old Machine Shop area of che Service Building and includes the following equipment. f i

rmergency ventilation Filter System I 1) i I

) The filter is sized to maintain a positive pressure in the TSC, [

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llVAC room and Filter room of 1/8" 11 0 during an emergency.

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2) IIVAC Systemn l

i Two completely separato and independent ilVAC systems are provided ,

for the TSC. j i

l The main TLC llVAC system consints of an air-cooled condensing unit, '

3 air handling unit with filter, cooling coil, heating coil, supply fan, supply and return air grilles and related ductwork.

I 1 The second IIVAC system handles the load associated with the pCS 3

{ mainframe.  !

During normal operation, the air handlers and exhaust nystem are i nized to maintain a slight positivo pressure in the TSC,llVAC room, i and the Filter room.

I j 3)  !!nttcry Room llVAC ,

l The Battery Room is air conditioned by a sing 1c-zone air handling i unit with an air-cooled condensing unit located on the roof of the i Service !!uilding.

j  !!ydrogen removal in accomplished by a amn11 exhaust fan and wall l louver. )

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! 4) Mineellaneous Exhaust and Ventilation i

The UPS aren in ventilated by a wn11 mounted louver and a roof exhaust fan.

The kitchen and toilet rooms nre being ventilated by an exhaust fan.

Additionni facilition required for the TSC have been installed under work covered by the following other DCps: 81-15, 81-16, 81-18, 81-19.

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! TECilNICAL SUPPORT ChNTER FACILITIES INSTALLATION i

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Summary of Safety Analysis (cont.)

i The TSC facilities installation has no effect on station operation or the operation of safety-related equipment. All TSC facilities are independent of the existing plant except for the TSC smoke and filter l

unit heat detection systems and filter unit controls. The TSC smoke and i filter unit heat detection systems are int 4 grated with the existing multiplexed smoke detection system and the filter unit actuates on a i Unit 1 or 2 safety injection signal. The above-mentioned system i interfaces do not affect the present operation of either the existing j smoke detection system or the safety injection instrumentation systems.

i I The Design Change does not involve modifications to safety reinted  !

! equipment previously evaluated in the UFSAR, systems that may in any way l j create accidents or malfunctions of a different type than previoisly I j evaluated in tha UFSAR nor systems whose margin of safety is defined in  !

the Technical Specifications.

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DC 84-52 UNIT 1 REGULATORY GUIDE 1.97 MAIN STEAM SAFETY RELIEF VALVE POSITION INDICATION Description The Main Steam Safety Valves (SVs), the Steam Atmospheric Dump Valves (PCVs) and Decay Heat Release Valve (HCV) did not have position indication as required by Regulatory Guide 1.97, Revision 3

" Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident".

The position of 15 Main Steam Safety Valves (SVs), 3 Steam Atmospheric Dump Valves (PCVs) and the Decay Heat Release Valve (HCV) which are located in the Main Steam Valve House (MSVH) at elevation 312'-0" are now monitored by Fluid Components Inc. temperature flow probes mounted in the discharge piping of the above valves.

Summary of Safety Analysis The additional flow sensors and flow transmitters are environmentally qualified, provide reliable indication for the Control Room Operator, and their installation does not effect safety equipment or operation of safety systems. l The additional flow eleuents that are installed in the Q. A. Category II i piping are environmentally qualified and provide indication of the SVs, ,

PCVs, and HCV position before and after a denign basis event. '

The additional flow sensors are qualified instruments which provide the Operator with reliable indication of the SV, PCV, and HCV position to assess the operation of safety systems and do not change safety limits defined in the Technical Specifications. The additional flow element and transmitter also provide the operator with reliable indication of steam flow to the turbine-driven auxiliary feedwater pump to assess the operation of thin nafety system and do not change safety limits defined in the Technical Speciffentions.

DC 84-54 UNIT 2 REGULATORY GUIDE 1.97 EMERGENCY VENTILATION DAMPER POSITION INDICATION Description '

Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions P ring and Following an Accident" describes an acceptable method for womplying with the NRC's regulations to provide instrumentation to monitor plant variables and systems during and following an accident to determine if the plant safety functions are being performed. The instrumentation must meet specific environmental, seismic, range, redundancy, power source and display requirements.

1 On January 31, 1984, the Virginia Electric and Power Company submitted I to the NRC under letter #054, the results of a comparison of the existing instrumentation and Reg. Guide 1.97, Revision 3 for North Anna l Units 1 and 2. Emergency Ventilation Damper Position was identified as

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a " Type D Variable" (Item D-34) and consequently inadequate to meet Reg.

Guide 1.97 for indication of damper status.

The following modifications were made to provide position indication of emergency ventilation dampers and satisfy R. G. 1.97 requirements:

MOV's-HV-204-1/2 were supplied with spara limit switches in the MOV Operator supplied by Rotork. Also a set of limit switches was wired to indicator lights on the Ventilation Panel in the Control Room. Position indication was utilized from the spare limit switches and wired into the 4

Validyne Data Acquisition System.

l i A0D's-HV-228-1/2/3/4 were supplied without limit switches.

Environmentally qualiffed limit switches were mounted such that the i original qualification and operation of the damper is not degraded. The

! Ifmit switches (open and closed) were wired into the Velidyne Data Acquisition System for position indication.

i l Summary of Safety Analysis t The Emergency Ventilation Damper Position Indication modification is

designed consistent with codes and standards of the existing systems.

This modification does not change the characteristics of any system in 4

operation during reactor operation or during construction implementation of the modification. l l

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DC 84-56 UNIT 2 REGULATORY GUIDE 1.97 HPSI FLOW TRANSMITTER MODIFICATION Description Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident" describes an acceptable method for complying with the NRC's regulations to provide instrumentation to monitor plant variables and systems during and following an accident to determine if the plant safety functions are being performed. The instrumentation must meet specific environmental, seismic, range, redundancy, power source, and display requirements.

On January 31, 1984, the Virginia Electric and Power Company submitted to the NRC, under letter #054, the results of a comparison of the existing instrumentation and Reg. Guide 1.97 Revision 3 for North Anna Units 1 and 2. The High Pressure Safety Injection Flow Instrumentation was identified as a " Plant Specific. Type A" variable (Item A-8) and consequently inadequate to meet R.G. 1.97 in the areas of redundancy, power source, seismic and environmental qualification.

The existing total High Pressure Safety Injection transmitters FT-2940 and FT- 2943 are environmentally qualified and will be retained. New power supply cards were relocated in the Secondary Plant Process Racks j to provide the necessary electrical separation and seismic mounting for the components in the instrument loop.

1 Two new transmitters FT-2940-1 and FT-2943-1 (one for the hot and one for the cold leg total flow) were installed in the Auxiliary Building to provide a redundant channel to monitor HPSI flow. These new transmitters utilize the same orifices (FE-2940 and FE-2943) as the existing transmitters and utilize the spare tubing stubs presently capped on each flow element flange.

Summary of Safety Analysis:

4 The HPSI flow instrumentation modification is designed consistent with codes and standards of the existing system. This modification does not change the characteristics of any system in operation during reactor operation or during construction implementation of the modification, i

DC 84-60 UNIT 2 EEGULATORY GUIDE 1.97 PRESSURIZER LIQUID TEMPERATURE TRANSMITTER MODIFICATION Description Regulatory Guide 1.97, Revision 3, " Instrumentation for Light-Water

-Cooled Nuclear Power Plant and Environs Conditions During and following an Accident," (R.C.1.97) requires that variables. Pressurizer Liquid Temperature was identified as a " Type A" variables (Item A-12) and as such must meet the Category I re:;uirements of R.G.

1.97. R.G. 1.97 requires that the entire instrument channel have an adequate range, be redundant and separated, have an adequate power supply, and be seismically and environmentally qualified.

There was only one Pressurizer Liquid Temperature channel, T-2453 and while the channel has the required range, it was not redundant, the wiring was color coded neutral, and it was not seismically or environmentally qualified.

The existing Pressurizer Liquid Temperature RTD, TE-2453, was removed and replaced with a new Weed dual element RTD. The RTD was seismically and environmentally qualified to IEEE 344-1975 and IEEE 323-1974.

The existing non safety related temperature channel T-2453 was deleted and two new redundant Pressurizer Liquid Temperature channels (T2453-1 red and T-2453-2 blue) were installed.

Summary of Safety Analysis:

The old Pressurizer Liquid Temperature channel T-2453 was a non safety-related channel cod as such, no credit was taken for it in any accident analyses previously evaluated in the UFSAR.

The replacement of the RTD with an environmentally and seismically qualified RTD reduces the probability of failure during normal and accident conditions.

The replacement of the' existing RTD with an environmentally and seismically qualified dual element RTD provides an increase in reliability for monitoring the, Pressurizer Liquid Temperature. This enhances the Control Room Operator's ability to assess the plant status during and after an accident.

The safety limits as defined in the Technical Specification were not affected by this design change.

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I DC 84-66 UNIT 2 MAIN STEAM VALVE MODIFICATION j Description 10CFR50, Appendix R, Section III, G.2, requires that certain means be provided to ensure the safe operation of redundant trains of systems i i

necessary to achieve and maintain safe shutdown conditions during a plant fire.

! The resolution for the problem of a fire in the Control Room or

! Emergency Switchgear Room that could adversely affect the control

' circuits for the Main Steam Isolation Valves (MSIVs), Turbine Steam Stop 1 Valves and Main Steam Dump Valves was to install a dedicated shutdown system for the MSIVs that will operate independent of the existing MSIV trip system.

j The solution for the problem of a fire in the Control Room or Emergency j Switchgear Room that could adversely affect the control circuits for the

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Steam Generator Power Operated Atmospheric Relief Valve problem was to install an instrument signal isolation switch in the Cable Vault and Tunnel to ensure the operators ability to close these valves in the event of a fire in the Emergency Switchgear Room or Control Room.

Summary of Safety Analysis This modification does not modify or alter the normal or accident operation functions of any existing control system or equipment.

i Instead it provides a means to prevent the loss of steam generator

, inventory following a fire in the Control Room or Emergency Switchgear

Room.

j The material used is environmentally qualified for the environment where it is installed. The material used is seismically supported where required to protect adjacent safety-related equipment from being damaged

, by this new equipment in the case of a seismic event.

1 This design enhances the ability of the Operator to maintain steam generator inventory by ensuring positive control of the Main Steam Isolation Valves and Steam Generator Power Operated Atmospheric Relief l Valves in the event of a fire in the Control Room or Emergency l Switchgear Room.

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f DC 85-19 UNIT 2 RTD BYPASS FLOW INDICATION Description Previously, the three RTD Bypass Flow indications consisted of Barton 288 differential pressure flow indicating switches FIC-0490, FIC-2491, and FIC-2492 with their associated orifice. The Barton indicating switches were located in containment and were used to provide a low flow alarm in the Control Room. When a low flow alarm occurred this indicated that the respective RTD temperature measurement could be erroneous. Verification of actual flow could only be made by entering containment and checking the local Barton indicator.

To provide remote verification of RTD bypass flow when an alarm occurs, the existing Barton differential pressure flow indicating switches were replaced with environmentally qualified Rosemount 1153 differential pressure transmitters are powered from a power supply installed in the Westinghouse process racks. A cabinet mounted indicator was provided and installed in a wall mounted box in the instrument rack room.

Summary of Safety Analysis The new transmitters are seismically supported and qualified for the environment in which they are used, and do not adversely affect safety related equipment.

This modification only provided for replacement of existing flow indicators with new flow transmitters and associated electronics to provide reliable flow indication.

This modification provided for more accurate and reliable remote measurement of RTD byphaa flow.

l' DC 84-50 UNIT 2 REGULATORY GUIDE 1.97 SERVICE WATER INSTRUMENTATION MODIFICATION Description Regulatory Guide 1.97 Rev. 3 requires the following parameters to be monitored at the plant Technical Support Center (TSC).

1) Auxiliary Service Water Pump discharge temperature (TE-SW211).

2) Component Cooling System chilled water supply temperature to the Containment Recirculating Fan Air Cooling coils (TE-CC264).

3) Coppeaent Cooling System Chilled Water return temperature to the Contsinment Recirculating Fan Air Cooling Coils (TE-CC267).

4) Component Cooling System Chilled Water return flow from the Containment Recirculating Fan Air Cooling Coils (FT-CC228)

This design change installed the necessary raceway and cabling to provide indication at the TSC via the Validyne Data Acquisition System.

Additionally, of the four parameters listed above, the Auxiliary Service Water Pump discharge temperature sensor must be environmentally qualified. This design change replaced TE-SW211 with an environmentally qualified RTD per R. G. 1.97 and IEEE 323-1974.

Summary of Safety Analysis The replacement of RTD TE-SW211 and the additional monitoring of FT-CC228, TE-CC264, TE-267, and TE-SW211 at the Technical Support Center does not constitute an "Unreviewed Safety Question" as defined in 10CFR50.59.

1. The implementation of this modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the Final Safety Anelysis Report. TE-SW211 will be upgraded by replacing it with an environmentally qualified RTD.

The additional monitoring of FT-CC228 TE-CC264, TE-CC267, and TE-SW211 at the Technical Support Center will enhance the ability of on-site personnel to evaluate the plant status during and after an accident.

2. The implementation of this modification does not create a possibility for an accident or a malfunction of a different type from any previously evaluated in the Final Safety Analysis Report.

This modification reduces the probability of failure of TE-SW211 which will increase the reliability of monitoring this parameter.

The additional monitoring of FT-CC228, TE-CC264, TE-267, and TE-SW211 will improve the ability of on-site personnel to evaluate the plant status during an emergency situation.

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l DC 84-50 UNIT 2 REGULATORY GUIDE 1.97 SERVICE WATER INSTRUMENTATION MODIFICATION (cont.)

3. The implementation of this modification does not reduce the margin of safety as defined in the basis of any Technical Specification.

The replacement of RTD TE-SW211 with a qualified RTD will provide the Control Room with reliable indication and does not change any safety limits defined in the Technical Specifications. Additional monitoring capability at the Technical Support Center will provide on-site personnel improved monitoring of emergency situations and does not change safety limits defined in the Technical Specifications, i

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DC 84-58

UNIT 2 REGULATORY GUIDE 1.97 CONTAINMENT ISOLATION VALVE POSITION INDICATION Description  !

Regulatory Guide 1.97 Rev. 3 requires that seismically and environmentally qualified position indication for Containment Isolation Valves must be provided to the Control Room Operator. In order to meet this criterion the following modifications were performed by this design change:

1) Accumulator Vent Line Control Valve (HCV-2936) was relocated above j the flood level in containment and environmentally and seismically l qualified limit switches were installed to provide control room

' valve position indication. The old switch in the control room was replaced with a seismically qualified switch that incorporates open

, and closed indicating lights.

2) A new Letdown Line Isolation valve, TV-2204A was installed inside containment in the Letdown Line between HCV-2200A, B, and C and

penetration 28. It assumes the Containment Isolation function of j HCV-2200 A, B and C, however, the Containment Isolation Phase A trip on HCV-2200 A, B and C will be retained in order to prevent the lifting of the relief valve RV-2203 upon a containment isolation signal. The new limit switches are' seismically and i

environmentally qualified. The new switch and indicating lights in the control room are seismically qualified in accordance with IEEE

344-1975 per Reg Guide 1.97.

3) Letdown Orifice Isolation Valves, HCV-2200A, B, and C. The supply voltage for the indicating light portion of the control circuits for HCV-2200 A, B, and C were changed from 125VDC to 120VAC. This will improve reliability by removing the potential of a ground on I. the 125VDC vital bus. While it does not prevent a ground on the 120VAC vital bus, this is not considered a problem because it is a grounded bus.

Summary of Safety Analysis

] 1. The implementation of this modification does not increase the j probability of occurrence or the consequences of an accident or i

malfunction of equipment important to safety and previously

evaluated in the Final Safety Analysis Report. The modifications j to the Letdown Line and the Accumulator Vent Line are designed l consistent with the Design Basis Criteria for Containment Isolation Valves and the requirements of Reg. Guide 1.97. The relocation of

( HCV-2936 does not functionally change the operation of the Safety

Injection System in any mode. The installation of TV-2204A does

!. not functionally change the operation of the Letdown System beyond requiring the opening the closing of TV-2204A at the same time as

TV-2204B ite redundant valve outside of containment. TV-2204A ,

1 ussumes the Inside Containment Isolation function of HCV-2200A, B, and C, however, these valves will still close on a Containment Isolation Signal. This modification does not affect any existing accident analyses.

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4 DC 84-58 UNIT 2 REGULATORY GUIDE 1.97 CONTAINMENT ISOLATION VALVE POSITION INDICATION l (cont.)

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2. The implementation of this modification does not crear.e a possibility for an accident or malfunction of a different type than

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any previously evaluated in the Final Safety Analysis Report. The m relocation of HCV-2936 does not functionally change the operation of the Safety Injection System in any mode. The installation of j TV-2204A does not functionally change the operation of the Letdown System beyond requiring the opening and closing of TV-2204A at the same time as TV-2204B, its redundant valve outside of containment.

t TV-2204A assumes the Inside Containment Isolation function of HCV-2200A, B, and C; however, these valves will still close on a Containment Isolation Signal. This modification does not affect any existing design analyses.

3) The implementation of this modification does not reduce the margin of safety as defined in the basis of any Technical Specification.

The safety limits as defined in the Technical Specifications are not changed.

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DC 84-62 i

UNIT 1 and 2 APPENDIX R-SERVICE WATER, COMPONENT COOLING WATER AND RESIDUAL HEAT REMOVAL

, PUMP CIRCUIT ISOLATION l

Description The Code of Federal Regulations, 10CFR50, Appendix R, Section III. L provides the requirements for alternative or dedicated shutdown capability that is provided for equipment required to operate to achieve cold shutdown conditions within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />.

The Appendix R review of the North Anna Units 1 and 2 Service Water, CCW, and RHR systems indicates that a fire in the Control Room could render any or all of these systems inoperative.

During post fire cooldown, one service water pump or one auxiliary service water pump is required for each reactor unit in operation.

l During p9st fire cooldown operation, one component cooling water pump is l required for each reactor unit.

i The Residual Heat Removal System transfers heat from the reactor coolant system to the component cooling water system to reduce the temperature I of the reactor coolant to the cold shutdown mode and maintains the reactor coolant temperature.

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To provide a means of isolating control room wiring and to provide a i means of operating Service Water pumps, Residual Heat Removal pumps and Component Cooling Water pumps remote from the control room, a transfer

switch and a control switch on each pump's breaker compartment at the i

switchgear was installed. The transfer is a 2-position selector switch j (LOCAL-REMOTE). The " LOCAL" position isolates all wiring between the 1 switchgear and the control room. The pump can be restored to the remote

, control mode by turning the transfer switch to the " REMOTE" position.

i The new local control switch operates the breaker only when the transfer

switch is in the " LOCAL" position.

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When the transfer switch at the switchgear is in the local position,

both the red and and green status lights in the Control Room for the i affected pump will be out. This will indicate to the operator that j control of that pump from the Control Room has been lost.

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! Summary of Safety Analysis i

1. The implementation of this modification does not increase the

probability of occurrence or the consequence of an accident previously evaluated in the Final Safety Analysis Report. This i modification isolates the control circuit wiring uin the Control  !

l Room and provides local control at the Switchgear for Service i

, Water, CCW, and RHR pumps. This modification will ensure that l operation of the affected pumps will be available in the event of a

! fire in the Control Room. This modification does not affect the. ,

l' safety operation of the equipment.

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l f DC 84-62 UNIT 1 and 2 APPENDIX R SERVICE WATER, COMPONENT COOLING WATER AND RESIDUAL HEAT REMOVAL PUMP CIRCUIT ISOLATION (cont.)

2. The implementation of this modification does not create a possibility for an accident different than any previously evaluated in the Final Safety Analysis Report. The modification adds isolation switches in the Emergency Switchgear room to disable the portion of the control circuit for the Service Water, CCW, and RHR

pumps routed in the Control Room and adds control switches to provide local control at the switchgear.

These changes will ensure that a fire in the Control Room will not affect the operability of the above equipment. New wiring and switches will be seismically qualified and seismically installed.

The addition of switches does not affect the original seismic qualification of the switchgear. This modification is consistent with station design criteria and all components used are qualified to the environment in which they were installed.

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3. The implementation of this modification does not reduce the margin of safety as defined in the basis of any Technical Specification.

The modification will ensure the Service Water, CCW, and RHR systems are available during a fire in the Control Room. This will result in continued operations necessary to achieve safe shutdown following a fire. Safety limits as defined in the Technical Specifications are not changed.

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DC 85-13 (UNIT 1)

DC 85-14 (UNIT 2)

FEEDWATER RECIRCULATION LINE LEAK MODIFICATIONS 4 Description Operational difficulties had been experienced with the feedwater recirculation system since initial plant startup. Pipe failures had occurred in areas downstream of the control valves, and weld-cracking was noted at piping connections at the condenser.

Failure of the feedwater recirculation piping was attributed to cavitation, erosion, and water hammer effects caused by the deterioration and poor performance of the flow control valves and restricting orifices in the system. Consequently, the valve and orifice sections were replaced by components of enhanced design. Also, a pipe support re-analysis was performed to ensure that the recirculation lines are adequately supported to absorb vibration and provide for any increased weight of new equipment. The piping layout was also redesigned to reduce the number of transitions.

Summary of Safety Analysis The replacement of the flow control valves and restricting orifices portions of the feedwater recirculation system is non-safety-related.

The work is being performed to enhance the operation of the feedwater recirculation system. Therefore, the Feedwater Recirculation Line Leak Modification as described herein does not constitute an "unreviewed safety question" as defined in 10CFR50.59, since it does not:

a) Increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the Updated Final Safety Analysis Report; this design change is non-safety-related and is being implemented to improve system performance, b) Create a possibility for an accident or malfunction of a different type than any evaluated previously in the Updated Final Safety Analysis Report; this design change will implement an enhanced feedwater recirculation system and will reduce the probability of system malfunction or failure.

c) Reduce the margin of safety as defined in the basis of any Technical Specification; the feedwater recirculation system is non-safety-related and is not addressed by the Technical Specifications.

DC 84-105 UNIT 1 and 2 PIPING PRESERVATION AUXILIARY I SERVICE WATER SUPPLY Description The service water system piping at North Anna Power Station has undergone severe, corrosive attack due to the aggressive nature of.the water and the presence of sulfate reducing bacteria. The corrosion is in the form of general wall thickness reduction and wide spread pitting, i Some small bore piping has required replacement due to through wall leaks or restricted flow. l l

This design change accomplished the cleaning and coating of portions of the Auxiliary Service Water Pump piping. Specifically, the lines that j were hydrolazed and coated are:

1 24"-WS-25-151-Q3 From valve MOV-SW-ll7 up to expansion joint in the 3

auxiliary service water pumphouse.

24"-WS-425-151-Q3 From valve MOV-SW-217 up to expansion joint in the auxiliary service water pumphouse.

The pipe cleaning was accomplished by a hydrolazing process. The majority of the inside surface, including pits, was cleaned to bare metal. Some tenacious modules of corrosion product remained. These l will be removed at a later date by chemical cleaning. Significant l amounts of water and corrosion products were generated from the j hydrolazing process; therefore, suitable temporary provisions were utilized for their collection and disposal.

Summary of Safety Analysis i

l The implementation of this modification does not:

1. increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the Updated Final Safety Analysis Report.

The auxiliary service water supply lines are isolated from the main system by double isolation valves. This double isolation capability was maintained during most of the completion of this design change. Since only one of the two auxiliary service water supply lines was out of service at any one time, the other line was i available to provide make-up to the service water reservoir.

2. create a possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis J Report.

The only modification created by this design change was addition of the internal pipe coating and replaced valves. The replacement check valves, gasketing and bolting materials used in this design

, change package are compatible (relative to design, quality and

functional requirements) with original equipment. Therefore, the j possibility for an accident or malfunction of equipment important to safety was not increased.

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3. reduce the margin of safety as defined in the basis of any Technical Specification, as the system was returned to its original configuration when the procedure is completed.

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DC 85-22 (UNIT 1)

DC 85-23 (UNIT 2)

SEQUENCE of EVENTS RECORDER REPLACEMENT Description The old station sequence of events recorders (Hathaway Model 500) had limited capacity, provided poor performance, and in some cases failed to record some vital trips. The station also experienced problems with the recorder / teletype interface cards and the printers which were not designed for constant use.

The recorder manufacturer no longer manufactured or stocked spare parts.

Parts replacement was extremely slow and expensive, adversely affecting reliability / availability of alarm recording during station operation.

The replacement sequence of events recorders were installed into existing annunciator cabinets and interface with existing inputs through internal wiring and jack connectors. They have a capacity expandable to 1,024 points of sequential memory, with one millisecond resolution, and are proven state-of-the-art products with high reliability.

The new printers are keyboard send-receive (KSR) data terminals.

Summary of Safety Analysis The implementation of the sequence of events recorder replacement modification does not increase the probability of occurrence or the consequences of an accident previously evaluated in the Final Safety Analysis Report.

This modification involved the replacement of the existing sequence of events recorders with more reliable models and does not change safety systems or the operation of safety-related components.

The implementation of this modification does not create a possibility for an accident or a malfunction of a different type than any previously evaluated in the Final Safety Analysis Report.

The sequence of events recorders replacement do not impact existing station alarm circuit wiring.

The replacement recorders provide more reliable and more extensive means to determine the sequence of alarms and events / conditions which lead to plant trips.

The implementation of this modification does not reduce the margin of safety as defined in the basis of any Technical Specification.

The replacement sequence of events recorder do not change safety limits.

DC 85-26 (UNIT 1)

DC 85-27 (UNIT 2)

TURBINE DRIVEN AUXILIARY FEEDWATER PUMP LUBE OIL RESERVOIR l LEVEL MONITORING AND ALARM SYSTEM 1

Description To furnish operators with sufficient turbine driven auxiliary feedwater pump status information, seismically qualified level switches were '

installed in the lube oil reservoirs. The level switches monitor lube oil levels and are furnished with two independent output contacts. On a falling level, the output contacts operate so as to energize annunciator windows for each unit in the Control Room (one Train A and the other Train B). The trouble alarms alert operators of the availability status of the turbine driven auxiliary feedwater pumps.

The level switches were installed in the existing " plugged" reservoir opening located on the top surface of the lube oil reservoir.

Summary of Safety Analysis The implementation of this modification does not increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety as previously evaluated in the Final Safety Analysis Report._ The switches are seismically qualified. The level switches provide information to the operator to monitor pump operation status. Installation of the level switches does not affect equipment or the operation of safety systems.

The implementation of this modification does not create a possibility for an accident or a malfunction of a different type than any previously evaluated. The construction and post-construction phases of this modification did not affect the turbine-driven auxiliary feedwater pump starting logic or operation criteria.

The implementation of this modification does not reduce the margin of safety as defined in the basis of any Technical Specification. The modification added seismically qualified level switches and did not change safety limits defined in the Technical Specification. This change was performed in accordance with the limiting conditions for availability of the turbine driven auxiliary feedwater pumps.

l DC 83-03 UNIT I and 2 REPLACEMENT OF EMERGENCY DIESEL 40X RELAYS Description The Gould ITE 40X relays that were originally installed in the Emergency Diesel Generator (EDG) under-excitation alarm circuits for lH, IJ, 2H, and 2J EDGs were subject to repeated failure. The 40/76 relay contacts were damaged by arcing. An engineering study found that the current draw of the 40X relay exceeded the 40/76 relay contact duty by approximately 650 percent. Carbon deposits and arcing damage of the 40/76 relay contact caused the 40X relay to chatter which ultimately led to 40X coil failure.

The Gould ITE 40X relays in the EDG under-excitation alarm circuits in EDG's, lH, IJ, 2H, and 2J were replaced with GE type HGA relays.

Replacing the existing Gould ITE 40X relays with GE type HGA relays will lower the contact duty of the 40/76 relays to within their limits. The replacement of the Gould ITE 40X relays do not affect normal station operation or the operation of any safety related equipment or systems, since they are used for alarm purposes only. These relays are located in each of the diesel room local annunciator panels.

Summary of Safety Analysis i The replacement of the Gould ITE 40X relays does not constitute an "unreviewed safety question" as defined in 10CFR50.59. This modification does not:

1. Increase the probability of occurrence or consequences of an accident or malfunction of equipment important to safety and previously evaluated in the UFSAR since the relay replacement will not affect the operation of any equipment or systems required to mitigate a Design Basis Accident. The relay is for under excitation alarm purposes only,

2. Create the possibility of an accident or malfunction of a different type than evaluated in the UFSAR,

3. Reduce the margin of safety as defined in the basis for any Technical Specification. This addition does not alter the basis for any Technical Specification because no safety related component or system, operation or effect is changed. The original relays were replaced by new relays to enhance system reliability.

DC 83-31 UNIT 2 APPENDIX "R" EXCORE NEUTRON FLUX MONITOR SYSTEM MODIFICATIONS Description 10CFR50 Appendix "R" requires that at least one train of excore neutron flux monitoring be available following a fire to achieve a cold shutdown conditions.

Regulatory Guide 1.97, Rev. 3 requires that excore neutron flux monitoring instrumentation should be ca sidered Class 1E, Seismic Category I, and should provide redundant indication of neutron flux over a range of 10E-6 to 100% of full reactor power.

The system provides overlapping source and power range neutron flux indication in the Technical Support Center (TSC), Control Room, and in the Fuel Building at the Remote Monitoring Panel. Indication in the TSC and Control Room is provided for post-accident monitoring and is designated as Class IE. Indication at the Remote Monitoring Panel is for use following a fire in either the Control Room, Emergency Switchgear Room, or the Cable Tunnel and Vault and is designated as non-class 1E.

Summary of Safety Analysis The Excore Neutron Flux Monitor System provides information to be used by the operators following an accident or a major fire. There are no possible accidents or malfunctions of equipment important to safety which would have resulted in a higher probability of occurrence or increased consequences as a result of this modification. The system is supplied by Class 1E power sources (120VAC vital buses). Non-safety related portions of the system are isolated from safety-related portions by the Gamma-Metrics Signal Processors. Safety-related portions are seismically qualified and supported.

The layout, function, and design of the system ensures physical and functional isolation from other equipment which is important to safety.

Therefore, there is no possibility of an accident or malfunction of a different type than any already evaluated.

The Excore Neutron Flux Monitor System has no effect on the margin of safety defined in the basis of any Technical Specification. The flux detectors are located in wells in the Neutron Shield Tank where they have no effect on other instrumentation systems found in the Technical Specifications.

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• DC 85-15 UNIT 1 and 2 TURB1NE BUILDING ROOF FIRE PROTECTION Description This design change addressed concerns by American Nuclear Insurers (ANI) due to an ANI inspection of North Anna following the main transformer fire on July 3, 1981.

ANI Recommendation 81-7 was concerned with providing additional fire protection capabilities on the turbine building roof. It stated that the turbine building roof should have an additional access method in the event of an emergency where access to the roof is necessary. In additien, the roof should be equipped with a fire water system standpipe protected from freezing and properly equipped hose houses. The fire hoses should be of sufficient length to allow deployment to any part of the roof from the standpipe.

The recommendation for additional access to the turbine building roof was resolved using Engineering Work Request 83-508. A new open exterior ladder was installed at the south-southwest corner of the turbine building.

The new standpipe is essentially a roof hydrant. The standpipe has two 2-1/2 inch outlets. The standpipe is equipped with the fire hoses (2-1/2" and 1-1/2") gated wyes, fire hose nozzles and wrenches. The hose provided is adequate to provide two hose streams on any part of the roof. The standpipe is housed in a hydrant house similar to the - type used for the ground hydrants and is located approximately in the middle of the turbine building on the south wall. The area of protection for the ground hydrants and is located approximately in the middle of the turbine building on the south wall. The area of protection for the standpipe is equivalent to the area of protection for a fire hydrant installed on the ground. The furthest distance from the standpipe to any location on the roof is 400 feet. The standpipe is sized to provide 200 gpm at the standpipe outlets at the required minimum pressure of 65 psi.

Safety railings were installed along the edges of the turbine building roof for the safety of fire fighters.

Summary of Safety Analysis This design change does not constitute an "Unreviewed Safety Question" as defined in 10CFR50.59 since it does not:

1. Increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the Updated Final Safety Analysis Report.

This modification would decrease the consequences of a fire on the turbine building roof.

DC 85-15 UNIT 1 and 2 TURBINE BUILDING ROOF FIRE PROTECTION (cont.) j

2. Create the possibility for an accident or nalfunction of a different type than previously evaluated in the Updated Final Safety Analysis Report. The addition of a standpipe will only increase the capabilities of fire protection on the turbine building roof.

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3. Reduce the margin of safety as defined in the basis of any Technical Specification. The design change does not impact any accident analysis and will increase the number of standpipes at the station and therefore increase the fire protection capability.

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DC 85-30 UNIT 2 REPLACEMENT OF STATION BATTERIES Description I The originally installed 125V DC Station Batteries (excepting 2-IV) were approaching the end of service life and required replacement. The original DC load calculations and battery sizing requirements of IEEE STD-485-1978 were reviewed for recommended design margin. New DC load calculations were performed, in accordance with IEEE STD -485-1983, for battery sizing to meet the present bus loads, including loads added since installation. The new calculations indicated that larger capacity batteries were required to provide an adequate safety margin.

New battery charger calculations indicated that the existing battery chargers are adequate and are not required to be changed. New battery conductor calculations indicated that the existing battery conductors (3-250 McM) are adequate and do not require replacement.

The design change installed the new batteries (with a larger capacity),

battery racks (2 tier), inter-rack cabling and conduit in Battery Rooms 2-1, 21I, and 2-III. The new batteries are Exide "2GN23" and consist of thirty two (32) cell units per battery room. The new batteries provide backup capacity needed for the operation of vital safety and nonsafety-related equipment necessary for the proper and safe operation of . the Station. An "Unreviewed Safety Question," as defined in 10CFR50.59, does not exist because:

A. The implementation of this modification does not increase the probability of occurrence of an accident or malfunction of equipment important to safety and previously evaluated in the Final Safety Analysis Report. This modification replaces existing Station Batteries in Battery Rooms 2-I, 2II, and 2-III with new and larger capacity batteries which will coordinate with the balance of the present 125V DC system.

The new batteries will be designed, fabricated, and installed to meet or exceed the requirement sections of the original Design Basis Document.

B. The implementation of this modification does not create a possibility for an accident or malfunction of a different type than any evaluated previously in the Final Safety Analysis Report. The materials used in the modifications are totally compatible (relative to design, quality, and functional requirements) with original equipment. Therefore, the possibility for an accident or malfunction of a different type than previously evaluated does not exist.

C. The implementation of this modification does not reduce the margin of safety as defined in the basis of any Technical Specification. New larger capacity batteries will be installed in Battery Rooms 1-I, 211, and 2-III enhancing the present 125 V DC system. Safety limits as defined in the Technical Specifications are not changed.

DC 85-33 UNIT 1 AND 2 CAS STRIPPER MOISTURE SEPARATION MODIFICATION - ADDITION OF KNOCK 0UT DRUM Description The Boron Recovery System includes the Gas Stripper Subsystem. The function of the Gas Stripper subsystem is to remove dissolved gases from the primary coolant letdown to the Boron Recovery system and to transfer them to the Gaseous Waste system or recycle them to the CVCS Volume Control Tanks. In the past, excessive water was being carried over from the gas strippers through the stripper vent chillers into the suction line of the stripper gas compressors. The excessive water in the suction line resulted in ruptures of the compressor seals and head gaskets. A reduction in the water carryover to the stripper gas compressors was accomplished by installing a knockout drum in the suction line to the stripper gas compressors. The knockout drum serves as a collection point for water and a surge volume for gas. Valving is provided to allow bypassing of the knockout drum, as well as, for draining the water accumulated in the drum to the suction lines of the "A" and "B" gas stripper discharge pumps. The drum is drained manually.

Level and pressure indicators are provided for draining operations.

Summary of Safety Analysis The modification has been reviewed pursuant to 10CFR50.59 and has been deemed not to involve an unreviewed safety question. The implementation of this modification does not:

1. Increase the probability of occurrence or the consequences of an accident or malfunction of equipment important to safety and previously evaluated in the Updated Final Safety Analysis Report.

The safety function of the gas stripper subsystem is to provide a pressure boundary for containing radioactive water and gas. There is no change since the added piping and fittings meet or exceed the t

design code requirements used during the original installation.

2. Create a possibility for an accident or a malfunction of a different type than any evaluated in the Updated Final Safety Analysis Report. The addition of the knockout drum should reduce the failures experienced with the stripper gas compressors due to moisture carry-over and enhance the operation of the Gaseous Waste System.

3. Reduce the margin of safety as defined in the basis of any Technical Specification since no Technical Specifications are affected.

DC 85-36 UNIT 2 CONDENSER MODIFICATIONS IMPINGEMENT GRATING INSTALLATION Description Foreign objects left from repair and maintenance operations on steam ,

lines, turbine and condenser have been found in the tube bundles and hotwells of the main condenser. These objects including nails, welding rod stubs, bolts, and wire wrap, can become missiles which cause impingement damage to the condenser tube bundles.

Turbine extraction steam pipfug and the 5th and 6th Point feedwater Heaters located inside the main condenser shells are covered with sheet metal lagging to increase turbine and feedwater heater thermal efficiency. Lagging failures have resulted in pieces of lagging impinging on and penetrating condenser tubes or becoming debris in the tube bundle. Lagging failures have resulted in voids in the thermal lagging surface and therefore degrade the thermal protection / efficiency of the heaters and extraction piping.and extraction piping.

The resolution was to repair the lagging on the 5th and 6th Point Feedwater Heaters and turbine extraction lines inside the condenser j shells and install impingement grating to protect the condenser tubes j from missiles and debris.

l Summary of Safety Analysis Modifications to the condenser are classified as Category II and are not safety related. Modifications were completed during a unit shutdown and, therefore, do not affect safe operation of the unit. This design change does not affect condenser / turbine or unit operation after installation.

The main condenser is not required to prevent or mitigate the consequences of any reactor coolant.

The Technical Specifications do not provide any limiting conditions for operation or surveillance requirements for the main condenser.

Administrative Controls (Section 6.8.4.c) for Secondary Water Chemistry are not affected by this design change.

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